The operating environment within a nuclear reactor, including a pressurized water reactor (PWR) and a boiling water reactor (BWR) is particularly hostile. A considerable effort has been expended in the nuclear reactor industry to arrive at materials which are able to withstand the combination of mechanical, thermal, chemical (corrosion) and radiation effects encountered in that environment. At the present time, only a few types of zirconium-based alloys are considered to be acceptable. Those alloys are generally identified as Zircaloy materials. The Zircaloy materials are used for nuclear fuel cladding tubes, spacer elements and channels within the reactor.
As a result of experience with long term operation and multiple reloads of nuclear fuel elements, it has been found that certain operating conditions arise which tend to reduce the energy output per unit of fuel ("burn-up") obtainable and thereby affect operating costs and efficiencies in an undesirable manner. For example, during the operation of nuclear reactors, metal debris which may be present in the reactor can be carried by the cooling water and can impact upon fuel assembly components. The repeated interaction of such debris and the fuel assembly components (such as fuel cladding tubes, channels or spacer elements) can result in fretting (rubbing) damage to the components.
While the Zircaloy materials gradually have been optimized with respect to corrosion resistance requirements within a reactor, the fretting wear resistance of Zircaloy, as well as resistance to combined effects of fretting wear and subsequent corrosion have not been optimized. The need to improve fretting wear resistance should not result in any undesirable compromise with respect to corrosion resistance.
Zircaloy materials until relatively recently have been treated prior to insertion into a reactor by autoclaving techniques to apply a relatively thin coating (0.5 microns) of oxide material to improve their general operational characteristics. Such an oxide coating has not been found to be resistant to fretting wear or fretting induced corrosion but rather has been found to be subject to being damaged or worn away by the fretting action of the debris. Thereafter, fretting corrosion will occur at the fretting site in the area where the oxide layer has been removed. The corrosion layer which forms is also susceptible to debris fretting wear and will be removed by action of the water and debris. Eventually, after successive cycles of wear and corrosion occur, a hole ultimately can be produced in the base metal itself. In the case of fuel cladding, such a hole will result in the unwanted release of radioactive material and radiation into the cooling water, and if it is in excess of reactor operating limits, will require an untimely shutdown of the reactor for replacement of fuel elements.
One approach to avoiding such problems is to improve the wear resistance of the Zircaloy fuel assembly components, especially at their lowermost ends where debris is most often present.
Outside of the field of nuclear reactors, it has been proposed that layered structures incorporating whiskers of nitrides, carbides or carbonitrides into Group IVb metals, which include zirconium, will provide a hardened surface condition. (See, e.g., U.S. Pat. Nos. 4,915,734; 4,900,525; and 4,892,792.) Furthermore, dispersions of hard substances in a binder metal, such as zirconium oxide dispersed in iron, cobalt or nickel (see U.S. Pat. No. 4,728,579) have been described as providing improved wear resistance for cutting tools. In addition, in the field of cutting tools, it has been observed (see U.S. Pat. No. 3,955,038) that, before applying an oxide coating such as zirconium oxide to a binder metal, imposition of an intermediate layer such as a carbide or a nitride of a metal in the fourth to sixth subgroups of the periodic system (including zirconium) may impede undesirable diffusion of metal from the substrate into the formed oxide layer.
It is understood that the foregoing structures are formed by processes which require temperatures that are incompatible with maintaining the metallurgical state of Zircaloy components to be used in a reactor. That is, such Zircaloy components typically are heat treated to produce particular grain structures and stress relieved conditions in the finished product.
In order to preserve the desired metallurgical structure and properties, it is necessary that any additional wear resistant layer be applied utilizing temperatures that are below temperatures at which the desired properties will be changed.
For example, in the case of stress relieved cladding of a type used in pressurized water reactors, post annealing processing temperatures should be maintained below about 500.degree. C. In the case of cladding, spacers or channels which have been treated to produce a recrystallized condition (as typically used in boiling water reactors), post annealing processing temperatures should be maintained below about 700.degree. C. and, in some cases, below about 600.degree. C. in order to avoid undesired metallurgical changes in the respective components.